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% Doug Wright
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% COK-2001-600 
% 11-304 Physics concepts such as hydrodynamics, photon transport,
% neutronics, fission, fusion, etc., when no classified information
% or association is revealed.

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\section{User Manual}

This section describes how to use this software library to accurately simulate neutron and gamma-ray emission from individual fission reactions.
The latest version of the library can be downloaded from \httpnuclear. Consult the file \texttt{Release\_notes.txt} in the software release to see the compatibility of this library with specific versions of {\tt MCNPX}, {\tt MCNP6}, and {\tt Geant4}.  Earlier versions of the library are distributed in the public release of {\tt MCNPX 2.7.0}~\cite{MCNPX} and {\tt Geant4}~\cite{Geant1,Geant2}.

The following sections describe how to run this library with {\tt MCNPX}/{\tt MCNP6} (Section~\ref{sec:mcnpx}) and {\tt Geant4} (Section~\ref{sec:geant4}), while Section~\ref{sec:api} describes the programmer's interface. 
For examples of creating a stand-alone executable with the programmer's interface, consult the directory \texttt{regr} in the software release.

\subsection{Limitations of the fission library~\label{Limitations of the fission library}}

The range of neutron energies for which induced fission neutron 
multiplicity data are available in the literature spans the range 
from 0 to 10~MeV, to which corresponds a range of $\bar{\nu}$ 
values.  The sampling of number of neutrons per fission is based on 
either the incident neutron energy or the $\bar{\nu}$ corresponding to 
that energy (depending on the option selected in \textit{setnudist}). 

When sampling is based on $\bar{\nu}$ (the default), 
and the $\bar{\nu}$ is in the range for which we have multiplicity 
data from the literature, that data is used. Outside that 
range, the Terrell approximation is used.

If the user selects the option to sample based on energy, and
the energy is within the range for which we have multiplicity 
data from the literature, that data is used.  If the energy is above 10 MeV, then 
the 10 MeV data is used. As this will be inaccurate as the energy becomes much higher
than 10 MeV, the user should select sampling based on $\bar{\nu}$ in this case. 
The same considerations apply for photofission.

In the case of spontaneous fission, data is only available 
for the following isotopes: $^{232}$Th, $^{232}$U, 
$^{233}$U, $^{234}$U, $^{235}$U, $^{236}$U, $^{238}$U, 
$^{237}$Np, $^{238}$Pu, $^{239}$Pu, $^{240}$Pu, $^{241}$Pu, 
$^{242}$Pu, $^{241}$Am, $^{242}$Cm, $^{244}$Cm, $^{249}$Bk, 
and $^{252}$Cf. 
The Monte-Carlo codes {\tt MCNPX} and {\tt Geant4} do not emit any particles if a different spontaneous fission 
isotope is specified.

\input{mcnpx.tex}
\input{geant4.tex}
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\input{interface.tex}
